This invention relates generally to nuclear reactors and more particularly, to systems and methods for computing neutron and gamma fluence in a nuclear reactor.
A reactor pressure vessel (RPV) of a boiling water reactor (BWR) typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A core assembly is contained within the RPV and includes the core support plate, fuel bundles, control rod blades and a top guide. A core shroud typically surrounds the core assembly and is supported by a shroud support structure. Particularly, the shroud has a generally cylindrical shape and surrounds both the core plate and the top guide. There is a space or annulus located between the cylindrical reactor pressure vessel and the cylindrically shaped shroud.
The core of the reactor includes an array of fuel bundles with square cross section. The fuel bundles are supported from below by a fuel support. Each fuel support supports a group of four fuel bundles, with the exception of the peripheral fuel supports which support a single fuel bundle. The thermal power generated in the core can be decreased by inserting control rods into the core, and the generated thermal power can be increased by retracting control rods from the core. In some BWR's, the control rods have a cruciform cross section with blades that can be inserted between the fuel bundles of a group of four.
Internal structures of operating BWRs are susceptible to various corrosive and cracking processes. Stress corrosion cracking (SCC) is one known phenomenon occurring in reactor components, such as structural members, piping, control rod guide tubes, fasteners, and welds, exposed to high temperature water. The reactor components are subject to a variety of stresses associated with, for example, differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stresses from welding, cold working and other inhomogeneous metal treatments. In addition, water chemistry, welding, heat treatment and radiation can increase the susceptibility of metal in a component to SCC.
It has been recognized that radiation creates oxygen and hydrogen peroxide via radiolysis, and that these chemical species significantly increase the electrochemical corrosion potential (ECP) throughout the primary circuit. This, in turn, assists stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC) of internal components of RPVs. Neutron radiation is especially efficient at creating oxygen and hydrogen peroxide in the reactor water within the core, while gamma radiation tends to promote the reduction of these species by hydrogen within the downcomer.
As nuclear power plants age, there is a growing need to better understand age-related degradation of the reactor pressure vessel and its internal components. At present, radiation dose is estimated by deterministic neutron transport codes. The accuracy of these estimates is not considered high enough due to inherent approximations in the geometric model of the system and the nuclear cross-section database. Accuracy is a particular problem in RPV regions where the dose gradient is very high, for example, outside the BWR core. There is presently a wide gap between the accuracy of present methods available compared to the desired accuracy to reliably evaluate continued degradation over time. The gap becomes even greater when continued operation for extended plant lifetimes is considered.